• Journal of the Korea Institute of Information Security & Cryptology
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• v.25 no.2
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• pp.363-374
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• 2015

Instrumentation and control(I&C) system has been mainly designed and operated based on analog technologies in existing Nuclear Power Plants(NPPs). However, As the development of Information Technology(IT), digital technologies are gradually being adopted in newly built NPPs. I&C System based on digital technologies has many advantages but it is vulnerable to cyber threat. For this reason, cyber threat adversely affects on safety and reliability of I&C system as well as the entire NPPs. Therefore, the software equipped to NPPs should be developed with cyber security attributes from the initiation phase of software development life cycle. Moreover through cyber security assessment, the degree of confidence concerning cyber security should be measured and if managerial, technical and operational work measures are implemented as intended should be reviewed in order to protect the I&C systems and information. Currently the overall cyber security program, including cyber security assessment, is not established on I&C systems. In this paper, we propose cyber security assessment methods in the Software Development Life Cycle by drawing cyber security activities and assessment items based on regulatory guides and standard technologies concerned with NPPs.

• Proceedings of the Korean Information Science Society Conference
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• 2002.10d
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• pp.19-21
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• 2002

Writing requirements in formal notation for a safety-critical system can improve software quality and reduce the errors that may arise later on in the software development life cycle. In this paper, we propose a formal specification approach used to describe the nuclear control system. The approach is based on the existing AECL approach that was the only formal specification technique applied to nuclear control systems in the past. Although the approach is AECL-based, the complex descriptions of certain requirements have been reduced by using different specification techniques. We discuss the differences and how the proposed approach provides not only specification but also verification environment.

• IE interfaces
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• v.21 no.1
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• pp.85-95
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• 2008

Reliability prediction serves many purposes during the life of a system, so several methods have been developed to predict the parts and systems reliability. MIL-HDBK-217F, among the those methods, has been widely used as a requisite tool for the reliability prediction which is applied to nuclear power plants and their safety regulations. This paper presents the reliability prediction for the DSP(Digital Signal Processor) module composed of three assemblies. One of the assemblies has a monitoring and self test function which is used to enhance the module reliability. The reliability of each assembly is predicted by MIL-HDBK-217F. Based on these predicted values, Markov modelling is finally used to predict the module reliability. Relax 7.7 software of Relax software corporation is used because it has many part libraries and easily handles Markov processes modelling.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2975-2981
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• 2020

Radiation damage is one of the aging important causes in nuclear reactors. Radiation damage causes changes in material properties. In this study, this effect has been evaluated and analyzed on the clad of the Tehran research reactor (TRR). A grade 6061 aluminum is used as a clad in the TRR. The MCNPX code is used to designate the most sensitive location of the reactor and calculate neutron flux distribution. Then, a software using FORTRAN language programming is developed to process the particle track (PTRAC) output file of the MCNPX code. The SRIM code is used here to calculate the rate of displacement per atom. Moreover, the SPECOMP and SPECTER codes are also applied to estimate the displacement rate and compared with the results attained using the SRIM code. The rate of displacement per atom by the SPECTER and SRIM codes have been obtained 2.54 × 10^-7 dpa/s and 2.44 × 10^-7 dpa/s (QD method), respectively. Also, the mechanical properties have been evaluated using the RCC-MRx code and have been compared with experimental results. Finally, the change in the matter specification has been analyzed as a function of time.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1842-1852
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• 2019

The AP1000 reactor coolant pump is a vertical shielded-mixed flow pump, is the most important coolant power supply and energy exchange equipment in nuclear reactor primary circuit system, whose steadystate and transient performance affect the safety of the whole nuclear island. Moreover, safety demonstration of reactor coolant pump is the most important step to judge whether it can be practiced, among which software simulation is the first step of theoretical verification. This paper mainly introduces the fluid-solid coupling simulation method applied to reactor coolant pump, studying the feasibility of simulation results based on workbench fluid-solid coupling technology. The study found that: for the unsteady calculations of the pure liquid media, the average head of the reactor coolant pump with bidirectional fluid-solid coupling decreases to a certain extent. And the coupling result is closer to the real experimental value. The large stress and deformation of rotor under different flow conditions are mainly distributed on impeller and idler, and the stress concentration mainly occurs at the junction of front cover plate and blade outlet. Among the factors that affect the dynamic stress change of rotor, the pressure load takes a dominant position.

• Proceedings of the KIEE Conference
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• 2007.04a
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• pp.143-145
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• 2007

The development of excitation control system (ECS) for emergency diesel generator in nuclear power plant is the replacement project of existing control system to resolve the maintenance problems caused by aging and obsolescence, The excitation control system is classified as a safety-related system. To guarantee the performance of developing excitation control system is equal to or higher than that of other systems, establishing the quality assurance scheme, doing software verification and validation activities, and planning equipment qualification. In this paper, we'd like to introduce the equipment qualification of excitation control system.

• Proceedings of the Korea Society for Simulation Conference
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• 1998.03a
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• pp.86-90
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• 1998

The safety and reliability of nuclear power plant operations relies heavily on the plant operators ability to respond to various emergency situations. It has become standard industry practice to utilize simulators to improve the safety and reliability of nuclear power plants operations. The simulators built for Younggwang#3,4, which is the basic model of the Korean Nuclear Power Plant design, has been developed precisely for this purpose. Dynamic Model and Display Model are developed under US3(UNIX Simulation Software Support System) environment in simulator for Younggwang#3,4. Since these two models are developed under each own operating system, it is necessary to develop a method for transporting data between these two systems. This paper descirves communication environment between Dynamic Model and Display Model, and addresses a file generation method for the Display Model, which will be necessary for designing MMI of MCR(Main Control Room) in the furture.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.373-377
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• 2023

The Monte Carlo (MC) method has become an indispensable part of the nuclear radiation research field. Several widely used and well-known MC packages were developed for simulation of radiation transport and interaction with matter. All these MC packages require users to prepare an input script. The input script can become lengthy for complex models. The process of preparing these input scripts is time-consuming and error-prone. In the present work, we have developed an open-source GUI computer program for modelling radiation transport and interaction in multi-segmented slab phantoms using grid-based system for the widely used PHITS MC package. The developed tools would be useful for future users of PHITS MC package and particularly inexperienced users. The present program is distributed under GPL license and all users can freely download, modify and redistribute the program without any restrictions.

• Nuclear Engineering and Technology
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• v.44 no.5
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• pp.471-482
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• 2012

To assess the risk of nuclear power plant operation and to determine the risk impact of digital systems, there is a need to quantitatively assess the reliability of the digital systems in a justifiable manner. The Probabilistic Risk Analysis (PRA) is a tool which can reveal shortcomings of the NPP design in general and PRA analysts have not had sufficient guiding principles in modelling particular digital components malfunctions. Currently digital I&C systems are mostly analyzed simply and conventionally in PRA, based on failure mode and effects analysis and fault tree modelling. More dynamic approaches are still in the trial stage and can be difficult to apply in full scale PRA-models. As basic events CPU failures, application software failures and common cause failures (CCF) between identical components are modelled.The primary goal is to model dependencies. However, it is not clear which failure modes or system parts CCF:s should be postulated for. A clear distinction can be made between the treatment of protection and control systems. There is a general consensus that protection systems shall be included in PRA, while control systems can be treated in a limited manner. OECD/NEA CSNI Working Group on Risk Assessment (WGRisk) has set up a task group, called DIGREL, to develop taxonomy of failure modes of digital components for the purposes of PRA. The taxonomy is aimed to be the basis of future modelling and quantification efforts. It will also help to define a structure for data collection and to review PRA studies.

• Proceedings of the Korean Information Science Society Conference
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• 2007.10b
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• pp.33-37
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• 2007

최근 디지털 컴퓨터와 정보처리기술의 발전과 더불어 원자력 발전소의 계측제어시스템과 같은 안전-필수 시스템에서도 디지털 기술을 채택하기 시작했다. 안전-필수 시스템에 사용되는 소프트웨어는 높은 신뢰도(dependability)가 요구된다. 소프트웨어의 신뢰도는 신뢰성(reliability), 안전성, 보안 등 다양한 속성들로 설명될 수 있다. 소프트웨어의 신뢰도 향상을 위한 프로세스는 결함예방프로세스, 결함허용프로세스, 결함제거프로세스 그리고 결함예측프로세스가 있으며 이들 프로세스는 소프트웨어 수명주기 초반부터 수행되어야 한다. 본 논문에서는 소프트웨어 신뢰도향상을 위한 신뢰도 프로세스 모델과 개발 단계별로 수행되어야 할 신뢰도 태스크를 제시한다.